Solution of neutron-transport multigroup equations system in subcritical systems

Abstract

Citation: Shamanin IV, Bedenko SV, Nesterov VN, Lutsik IO, Prets AA (2018) Solution of neutron-transport multigroup equations system in subcritical systems. Nuclear Energy and Technology 4(1): 79-85. https://doi. AbstractAn iteration method has been implemented to solve a neutron transport equation in a multigroup diffusion approximation. A thermoelectric generator containing plutonium dioxide, used as a source of thermal and electric power in spacecraft, was studied.Neutron yield and multigroup diffusion approximation data was used to obtain a continuous and group distribution of neutron flux density spectra in a subcritical multiplying system. Numerical multigroup approaches were employed using BNAB-78, a system of group constants, and other available evaluated nuclear data libraries (ROSFOND, BROND, BNAB, EXFOR and ENDSF).The functions of neutron distribution in the zero iteration for the system of multigroup equations were obtained by approximating an extensive list of calculated and experimental data offered by the EXFOR and ENDSF nuclear data libraries. The required neutronic functionals were obtained by solving a neutron transport equation in a 28-group diffusion approximation. The calculated data was verified. The approach used is more efficient in terms of computational efforts (the values of the neutron flux density fractions converge in the third iteration). The implemented technique can be used in nuclear and radiation safety problems. KeywordsSubcritical system, neutron distribution function, neutron transport, multigroup diffusion approximation. Research statusNeutron transport calculation methods are used in reactor physics, for process monitoring during nuclear fuel fabrication and reprocessing, to determine the irradiated fuel burnup, in biological shielding design, and to improve the radiation measurement procedures in the nuclear material accounting and control system.There are a lot of dedicated programs and codes that implement the Monte Carlo method based on selection of neutron interaction probabilities. The internationally recognized program packages include MMKKENO and MMKC (IPPE, Russia), MCU (Kurchatov Institute, Russia), MCNP (USA), KENO-3D (USA), MONK (Great Britain) and others. These programs make it possible to obtain highly accurate results thanks to a 3D geometry

Peculiarities of the radiation formation in dispersed microencapsulated nuclear fuel

“…For this reason, of greater interest is determination of the ratio of the quantity of the neutrons q an formed as a result of (a, n) reactions to the quantity of the neutrons resulting from spontaneous (q sf ) and induced (q f ) fission (in the event of large-size spent nuclear fuel (SNF) blocks) which is of interest both for calculating the neutron radiation dose rate during handling of such fuel Zabrodskaya 2005, Dulin andMatveyenko 2002) and in using the Rossi alpha method to determine the multiplication factor (k eff ) in subcritical systems (Dulin and Matveyenko 2002), such as the НGTRU fuel block (Shamanin et al 2018). This problem was examined rather thoroughly and solved in (Vlaskin et al 2015, Bulanenko 1979, Dulin and Zabrodskaya 2005, Dulin and Matveyenko 2002, Shamanin et al 2017, Bogatov et al 2015) and as part of other studies for fuel in the form of homogeneous dioxide. In the event when dispersed fuel and media containing heterogeneous inclusions of different configurations and sizes are considered, this problem is more difficult to solve and the method to calculate the quantitative and spectral compositions of the neutron source is of interest as such.…”